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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217D2721999-10-12012 October 1999 Safety Evaluation Supporting Amends 248 & 239 to Licenses DPR-77 & DPR-79,respectively ML20217B3651999-10-0606 October 1999 Safety Evaluation Supporting Amends 247 & 238 to Licenses DPR-77 & DPR-79,respectively ML20212J6311999-10-0101 October 1999 SER Accepting Request for Relief from ASME Boiler & Pressure Vessel Code,Section Xi,Requirements for Certain Inservice Insp at Plant,Unit 1 ML20212F0831999-09-23023 September 1999 Safety Evaluation Granting Relief from Certain Weld Insp at Sequoyah Nuclear Plant,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(ii) for Second 10-year ISI Interval ML20212F4761999-09-23023 September 1999 Safety Evaluation Supporting Amends 246 & 237 to Licenses DPR-77 & DPR-79,respectively ML20211N8891999-09-0707 September 1999 Safety Evaluation Supporting Amends 245 & 236 to Licenses DPR-77 & DPR-79,respectively ML20196J8521999-06-28028 June 1999 Safety Evaluation Authorizing Proposed Alternative to Use Iqis for Radiography Examinations as Provided for in ASME Section III,1992 Edition with 1993 Addenda,Pursuant to 10CFR50.55a(a)(3)(i) ML20206G3751999-05-0404 May 1999 Safety Evaluation Supporting Amends 244 & 235 to Licenses DPR-77 & DPR-79,respectively ML20205N0361999-04-12012 April 1999 Safety Evaluation Supporting Amend 234 to License DPR-79 ML20204E8211999-03-16016 March 1999 Safety Evaluation Supporting Amends 243 & 233 to Licenses DPR-77 & DPR-79,respectively ML20206U4331999-02-0909 February 1999 Safety Evaluation Supporting Amends 242 & 232 to Licenses DPR-77 & DPR-79,respectively ML20198C0211998-12-16016 December 1998 Safety Evaluation Supporting Amends 241 & 231 to Licenses DPR-77 & DPR-79,respectively ML20196J8021998-12-0707 December 1998 Safety Evaluation Supporting Amends 240 & 230 to Licenses DPR-77 & DPR-79,respectively ML20196B0231998-11-19019 November 1998 Safety Evaluation Supporting Amends 239 & 229 to Licenses DPR-77 & DPR-79,respectively ML20196C4091998-11-19019 November 1998 Safety Evaluation Supporting Amends 238 & 228 to Licenses DPR-77 & DPR-79,respectively ML20195G3271998-11-17017 November 1998 Safety Evaluation Supporting Amends 237 & 227 to Licenses DPR-77 & DPR-79,respectively ML20195H0831998-11-17017 November 1998 Safety Evaluation Supporting Amends 237 & 226 to Licenses DPR-77 & DPR-79,respectively ML20238F2961998-08-28028 August 1998 Safety Evaluation Supporting Amends 235 & 225 to Licenses DPR-77 & DPR-79,respectively ML20239A0631998-08-27027 August 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Sequoyah Nuclear Plant,Units 1 & 2 ML20236Y2091998-08-0707 August 1998 Safety Evaluation Accepting Relief Requests RP-03,RP-05, RP-07,RV-05 & RV-06 & Denying RV-07 & RV-08 ML20236S4081998-07-0101 July 1998 Safety Evaluation Supporting Amends 233 & 223 to Licenses DPR-77 & DPR-79,respectively ML20248L1961998-06-0808 June 1998 Safety Evaluation Supporting Amends 232 & 222 to Licenses DPR-77 & DPR-79,respectively ML20217K4471998-04-27027 April 1998 Safety Evaluation Supporting Requests for Relief 1-ISI-2 (Part 1),2-ISI-2 (Part 2),1-ISI-5,2-ISI-5,1-ISI-6,1-ISI-7, 2-ISI-7,ISPT-02,ISPT-04,ISPT-06,ISPT-07,ISPT-8,ISPT-01 & ISPT-05 ML20216E4701998-03-0909 March 1998 Safety Evaluation Approving Exemption from Updated FSAR Requirements of 10CFR50.71(e)(4) for Sequoyah Nuclear Plant,Units 1 & 2 ML20203K6811998-02-20020 February 1998 Safety Evaluation Supporting Amends 231 & 221 to Licenses DPR-77 & DPR-79,respectively ML20198P9171998-01-13013 January 1998 Safety Evaluation Supporting Amend 230 to License DPR-77 ML20217D7881997-09-29029 September 1997 Safety Evaluation Supporting Amends 229 & 220 to Licenses DPR-77 & DPR-79,respectively ML20211G2061997-09-23023 September 1997 Safety Evaluation Supporting Amends 228 & 219 to Licenses DPR-77 & DPR-79,respectively ML20210J5951997-08-12012 August 1997 Safety Evaluation Supporting Amends 227 & 218 to Licenses DPR-77 & DPR-79,respectively ML20151L7981997-07-14014 July 1997 Safety Evaluation Supporting Amends 226 & 217 to Licenses DPR-77 & DPR-79,respectively ML20148S0581997-07-0101 July 1997 Safety Evaluation Supporting Amends 225 & 216 to Licenses DPR-77 & DPR-79,respectively ML20140F0311997-06-10010 June 1997 Safety Evaluation Supporting Amends 224 & 215 to Licenses DPR-77 & DPR-79,respectively ML20138D2581997-04-28028 April 1997 Safety Evaluation Authorizing Licensee Proposed Alternative to Use 1989 Edition of ASME Boiler & Pressure Vessel Code, Section XI for Performance of Containment Repair & Replacement Activities Until 970909 ML20137Y8911997-04-21021 April 1997 Safety Evaluation Supporting Amends 223 & 214 to Licenses DPR-77 & DPR-79,respectively ML20137R6071997-04-0909 April 1997 Safety Evaluation Supporting Amends 222 & 213 to Licenses DPR-77 & DPR-79,respectively ML20101M5761996-04-0303 April 1996 Safety Evaluation Supporting Amend 211 to License DPR-79 ML20100N5711996-03-0404 March 1996 Safety Evaluation Supporting Amends 220 & 210 to Licenses DPR-77 & DPR-79,respectively ML20097D4191996-02-0707 February 1996 Safety Evaluation Supporting Amends 218 & 208 to Licenses NPF-77 & NPF-79,respectively ML20097F2211996-02-0505 February 1996 Safety Evaluation Supporting Amends 217 & 207 to Licenses DPR-77 & DPR-79,respectively ML20095F9891995-12-11011 December 1995 Safety Evaluation Supporting Amends 216 & 206 to Licenses DPR-77 & DPR-79,respectively ML20094N5341995-11-21021 November 1995 Safety Evaluation Supporting Amends 215 & 205 to Licenses DPR-77 & DPR-79,respectively ML20094D2451995-10-30030 October 1995 Safety Evaluation Supporting Amend 204 to License DPR-79 ML20093E1191995-10-11011 October 1995 Safety Evaluation Supporting Amend 214 to License DPR-77 ML20093B9471995-10-0404 October 1995 Safety Evaluation Supporting Amends 213 & 203 to Licenses DPR-77 & DPR-79,respectively ML20092N0711995-10-0202 October 1995 Safety Evaluation Supporting Amends 212 & 202 to Licenses DPR-77 & DPR-79,respectively ML20092H0811995-09-15015 September 1995 Safety Evaluation Supporting Amends 211 & 201 to Licenses DPR-77 & DPR-79,respectively ML20092G7071995-09-13013 September 1995 Safety Evaluation Supporting Amends 210 & 200 to Licenses DPR-77 & DPR-79 ML20087L4671995-08-22022 August 1995 Safety Evaluation Supporting Amends 207 & 197 to Licenses DPR-77 & DPR-79,respectively ML20086C2861995-06-29029 June 1995 Safety Evaluation Supporting Amends 205 & 195 to Licenses DPR-77 & DPR-79,respectively ML20085M1581995-06-14014 June 1995 Safety Evaluation Supporting Amends 204 & 194 to Licenses DPR-77 & DPR-79,respectively 1999-09-07
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEAR05000327/LER-1999-002-03, :on 990916,EDG Started as Result of Cable Being Damaged During Installation of Thermo-Lag for Kaowool Upgrade Project.Caused by Inadequate pre-job Briefing. Involved Individuals Were Counseled on Event1999-10-15015 October 1999
- on 990916,EDG Started as Result of Cable Being Damaged During Installation of Thermo-Lag for Kaowool Upgrade Project.Caused by Inadequate pre-job Briefing. Involved Individuals Were Counseled on Event
ML20217D2721999-10-12012 October 1999 Safety Evaluation Supporting Amends 248 & 239 to Licenses DPR-77 & DPR-79,respectively ML20217B3651999-10-0606 October 1999 Safety Evaluation Supporting Amends 247 & 238 to Licenses DPR-77 & DPR-79,respectively ML20212J6311999-10-0101 October 1999 SER Accepting Request for Relief from ASME Boiler & Pressure Vessel Code,Section Xi,Requirements for Certain Inservice Insp at Plant,Unit 1 ML20217G3721999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Sequoyah Nuclear Plant.With ML20212F4761999-09-23023 September 1999 Safety Evaluation Supporting Amends 246 & 237 to Licenses DPR-77 & DPR-79,respectively ML20212F0831999-09-23023 September 1999 Safety Evaluation Granting Relief from Certain Weld Insp at Sequoyah Nuclear Plant,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(ii) for Second 10-year ISI Interval ML20211N8891999-09-0707 September 1999 Safety Evaluation Supporting Amends 245 & 236 to Licenses DPR-77 & DPR-79,respectively ML20212C4761999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Sequoyah Nuclear Plant.With ML20212A1841999-08-25025 August 1999 Errata Pages for Rev 0 of WCAP-15224, Analysis of Capsule Y from TVA Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program ML20210L4361999-08-0202 August 1999 Cycle 9 12-Month SG Insp Rept ML20210L4451999-07-31031 July 1999 Unit-2 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept ML20216E3781999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20210G6631999-07-28028 July 1999 Cycle 9 90-Day ISI Summary Rept ML20209H3831999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Sequoyah Nuclear Plant.With ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20211F9031999-06-30030 June 1999 Cycle 9 Refueling Outage ML20196J8521999-06-28028 June 1999 Safety Evaluation Authorizing Proposed Alternative to Use Iqis for Radiography Examinations as Provided for in ASME Section III,1992 Edition with 1993 Addenda,Pursuant to 10CFR50.55a(a)(3)(i) ML20195K2951999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With 05000327/LER-1998-003-01, :on 981109,Vital Inverter 1-IV Tripped.Caused by Failed Oscillator Board with Bad Solder Joint,Attributed to Mfg Defect.Replaced Component & Returned Inverter to Operable Status1999-05-27027 May 1999
- on 981109,Vital Inverter 1-IV Tripped.Caused by Failed Oscillator Board with Bad Solder Joint,Attributed to Mfg Defect.Replaced Component & Returned Inverter to Operable Status
05000327/LER-1999-001-04, :on 990415,exceedance of AOT Occurred Due to Failure of Centrifugal Charging Pump.Rotating Element Was Replaced,Testing Was Completed & Pump Was Returned to Svc1999-05-11011 May 1999
- on 990415,exceedance of AOT Occurred Due to Failure of Centrifugal Charging Pump.Rotating Element Was Replaced,Testing Was Completed & Pump Was Returned to Svc
ML20206Q8951999-05-0505 May 1999 Rev 0 to L36 990415 802, COLR for Sequoyah Unit 2 Cycle 10 ML20206G3751999-05-0404 May 1999 Safety Evaluation Supporting Amends 244 & 235 to Licenses DPR-77 & DPR-79,respectively ML20206R5031999-04-30030 April 1999 Monthly Operating Repts for April 1999 for Sequoyah Units 1 & 2.With ML20205N0361999-04-12012 April 1999 Safety Evaluation Supporting Amend 234 to License DPR-79 ML20205P9811999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20204E8211999-03-16016 March 1999 Safety Evaluation Supporting Amends 243 & 233 to Licenses DPR-77 & DPR-79,respectively ML20205B6631999-02-28028 February 1999 Underground Storage Tank (Ust) Permanent Closure Rept, Sequoyah Nuclear Plant Security Backup DG Ust Sys ML20204C3111999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20203H7381999-02-18018 February 1999 Safety Evaluation of Topical Rept BAW-2328, Blended U Lead Test Assembly Design Rept. Rept Acceptable Subj to Listed Conditions ML20206U4331999-02-0909 February 1999 Safety Evaluation Supporting Amends 242 & 232 to Licenses DPR-77 & DPR-79,respectively ML20211A2021999-01-31031 January 1999 Non-proprietary TR WCAP-15129, Depth-Based SG Tube Repair Criteria for Axial PWSCC Dented TSP Intersections ML20198S7301998-12-31031 December 1998 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept ML20199G3641998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With 05000327/LER-1998-004-02, :on 981120,failure to Perform Surveillance within Required Time Interval,Was Determined.Caused by Leaking Vent Valve.Engineering Personnel Evaluated Alternative Methods for Performing Channel Check1998-12-21021 December 1998
- on 981120,failure to Perform Surveillance within Required Time Interval,Was Determined.Caused by Leaking Vent Valve.Engineering Personnel Evaluated Alternative Methods for Performing Channel Check
ML20198C0211998-12-16016 December 1998 Safety Evaluation Supporting Amends 241 & 231 to Licenses DPR-77 & DPR-79,respectively 05000327/LER-1998-003-04, :on 981109,automatic Reactor Trip Occurred. Caused by Component in Bridge Circuit of Vital Inverter Failed.Inverter Bridge Circuit Replaced1998-12-0909 December 1998
- on 981109,automatic Reactor Trip Occurred. Caused by Component in Bridge Circuit of Vital Inverter Failed.Inverter Bridge Circuit Replaced
ML20196J8021998-12-0707 December 1998 Safety Evaluation Supporting Amends 240 & 230 to Licenses DPR-77 & DPR-79,respectively ML20197J5621998-12-0303 December 1998 Unit 1 Cycle 9 90-Day ISI Summary Rept ML20197K1161998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20196C4091998-11-19019 November 1998 Safety Evaluation Supporting Amends 238 & 228 to Licenses DPR-77 & DPR-79,respectively ML20196B0231998-11-19019 November 1998 Safety Evaluation Supporting Amends 239 & 229 to Licenses DPR-77 & DPR-79,respectively ML20195H0831998-11-17017 November 1998 Safety Evaluation Supporting Amends 237 & 226 to Licenses DPR-77 & DPR-79,respectively ML20195G3271998-11-17017 November 1998 Safety Evaluation Supporting Amends 237 & 227 to Licenses DPR-77 & DPR-79,respectively 05000328/LER-1998-002-05, :on 981016,turbine Trip Occurred Followed by Reactor Trip.Caused by Generator Lockout.Mod Being Evaluated to Physically Isolate Relays from Vibration of Transformers & Adding Two of Two Logic for Actuation of Sudden Pressure1998-11-10010 November 1998
- on 981016,turbine Trip Occurred Followed by Reactor Trip.Caused by Generator Lockout.Mod Being Evaluated to Physically Isolate Relays from Vibration of Transformers & Adding Two of Two Logic for Actuation of Sudden Pressure
ML20195F8061998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Sequoyah Nuclear Plant.With ML20154H6091998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With 05000328/LER-1998-001-05, :on 980827,turbine Trip Occurred Followed by Reactor Trip.Caused by Failure of Sudden Pressure Relay on B Phase Main Transformer.Control Room Operators Responded & Removed,Inspected & Replaced Failed Relays1998-09-28028 September 1998
- on 980827,turbine Trip Occurred Followed by Reactor Trip.Caused by Failure of Sudden Pressure Relay on B Phase Main Transformer.Control Room Operators Responded & Removed,Inspected & Replaced Failed Relays
ML20154H6251998-09-17017 September 1998 Rev 0 to Sequoyah Nuclear Plant Unit 1 Cycle 10 Colr ML20153B0881998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Sequoyah Nuclear Plant.With 1999-09-07
[Table view] |
Text
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UNITED STATES O
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NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D.C. 2006H001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.238 TO FACILITY OPERATING LICENSE NO. DPR-77 AND AMENDMENT NO 228 TO FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328
1.0 INTRODUCTION
The Tennessee Valley Authority (TVA, the licensee) requested amendments to Operating Licenses DPR-77 and DPR-79 for Sequoyah Nuclear Plant (SON), Units 1 and 2, respectively, in a letter to the U.S. Nuclear Regulatory Commission (NRC) dated August 21,1996. The amendments would revise the SON Units 1 and 2 Technical Specifications (TS) to extend the applicability of the limiting condition of operation (LCO) for the condensate storage tank (CST) to Mode 4 when the steam generators are relied upon for heat removal. The licensee also proposed changes to associated action statements, surveillance requirements (SR), and TS bases.
2.0 BACKGROUND
in 1995, by license Amendment Nos. 206 and 196 for Units 1 and 2, respectively, the licensee revised TS 3.7.1.2, Auxiliary Feedwater System, to be consistent with the Westinghouse Standard Technical Specifications contained in NUREG-1431. This change expanded the mode applicability for the auxiliary feedwater (AFW) system's LCO from Modes 1,2, and 3, to include " MODE 4 when steam generator is relied upon for heat removal." At that time, the licensee did not recognize the need for a change to the mode of applicability for the CST which provides the water source for the AFW pumps. Presently, TS 3.7.1.3 requires the CST to be operable in Modes 1,2, and 3. This is not consistent with the TS required AFW operability in Mode 4. As an interim corrective action, the licensee is currently operating with administrative controls to ensure that CST is operable in all applicable modes when the AFW system is required to remain operable. The licensee has now proposed appropriate TS changes to correct the disparity between the CST and AFW TS.
9812O20047 981119
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DR ADOCK 05000327 PDH
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TVA proposes to modify TS 3.7.1.3, " Condensate Storage Tank" to:
(a) extend the mode of applicability to include " MODE 4 when steam generator is relied upon for heat removal." Current modes of applicability are Modes 1,2, and 3.
(b) revise Action (a) to increase the allowable action time for achieving hot shutdown to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from the current 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and to add the phrase "without reliance on steam 4'
generator for heat removal"to be consistent with the language in NUREG-1431.
(c) revise Action (b) to modify the phrase " Demonstrate the operability Essential Raw Cooling Water System as a backup supply to the auxiliary feedwater pumps.." to " Verify by administrative means operability of the Essential Raw Cooling Water System (ERCW) as a backup supply to the auxiliary feedwater pumps..." to be consistent with the language in 1
NUREG-1431.
(d) add the phrase "without reliance on steam generator for heat removal" to Action (b) to be consistent with the language in NUREG-1431.
l (e) delete the current SR 4.7.1.3.2 for verifying ER,CW operability every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and make the requirement a footnote to Action (b) instead.
(f) revise TS Bases 3/4.3.2 to provide guidance concerning the AFW suction pressure-low function as to the modes when this function is required to be operable.
(g) correct a typographical error in Action (b) by replacing "pmps" with " pumps."
j 3.0 EVALUATION Following a reactor trip, decay heat is dissipated by evaporating water in the steam generators and venting the steam either to the condensers or to the atmosphere. In such situations, steam generator water inventory must be maintained at a level sufficient to ensure adequate heat i
transfer and decay heat removal. The AFW system pumps deliver this emergency water supply to the steam generators. The AFW system provides emergency water to the steam generators until either normal feed water flow is established or the residual heat removal (RHR) system can assume the decay heat removal function. The primary sources of water for the AFW system pumps are the condensate storage tanks. On low suction pressure, the AFW pumps are designed to automatically swap to the ERCW system.
l SQN has two CSTs (one for each unit) that provide cooling water to the suction of the AFW pumps. TVA's proposed change to extend the mode of applicability for the CST to " MODE 4 i
when steam generator is relied upon for heat removal" would ensure that a source of water is available to the required AFW train until the RHR system can assume the decay heat removal function. This change is consistent with the mode requirements for AFW operability.
Therefore, the staff finds the proposed change to the CST operability acceptable.
,m,
1 3
Adding the phrase "without reliance on steam generator for heat removal" to current Actions (a) and (b) would be done for the pupoce of ensuring unit cooldown to RHR entry conditions without reliance on steam generators. The staff notes that this phase is consistent with the language in Section 3.7.6 of NUREG-1431 and is discussed in Bases Section B 3.7.6. The staff finds the proposed change to be consistent with the proposed mode of applicability for the CST and is, therefore, acceptable. The staff also finds the action time for achieving hot standby from 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable for transition from the steam generator mode of decay heat removal to RHR entry conditions and would achieve the required plant condition in an orderly manner without challenging other plant systems.
SR 4.7.1.3.2 for demonstrating operability of the ERCW system is redundant to TS 3.7.4 which requires two ERCW loops to be operable in Modes 1,2,3, and 4. SR 4.7.1.3.2 for demonstrating operability of the ERCW system is, therefore, not necessary and the staff finds deleting SR 4.7.1.3.2 to be acceptable.
The staff also finds the proposed TS changes discussed in items (c), (e), and (g) above to be editorial and, therefore, acceptable.
4.0 STATE CONSULTATION
in accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (61 FR 52967, dated October 9,1996). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety af the public Principal Contributor: L. Raghavan, NRR Date: tbveter 19, 1998 l
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